Systematic and exact scaling analysis of the single-phase natural circulation flow: The hydraulic similarity

For the study of the hydraulic similarity in a single-phase natural circulation loop, the integral momentum equation is non-dimensionalized with respect to the initial flow kinematic energy of reference section, without intuitively specifying any reference parameters. By this mean, a unique hydraulic time scale, characterizing the system hydraulic response, is identified along with two dimensionless physical numbers: the dimensionless flow resistance number and the dimensionless gravitational force number. From the integral momentum equation, the mass flow rate at steady state is also obtained. The identified dimensionless parameters are then applied to derive a set of scaling criteria for the design of a full-pressure reduced-size similar model for a PWR (Pressurized Water Reactor). For exact hydraulic similarity, it was found for the first time that the cross sectional area scaling ratio should be related to the axial length scaling ratio. In addition, it is also found out that the relative cross-sectional area ratio should be preserved in order to preserve the flow resistances. Moreover, the scaling ratio for the number of the U-tubes was found to be unity if exact hydraulic similarity is pursued for the whole system. Three sets of scaling criteria for the design of a full-pressure model for a PWR are summarized in a table for different application. The accuracy and applicability of this proposed scaling method is demonstrated by proposing a simple loop and a PWR-like system, by scaling down the systems to get two corresponding models with this proposed scaling methodology, and by comparing the model results with their corresponding prototype results. Furthermore, the method for the evaluation of both system-level and local hydraulic scaling distortions are addressed.     

Bifurcation and hysteresis phenomena in the two－phase natural circulation system

The static bifurcation of the two-phase natural circulation (TPNC) system was studied theoretically and numerically.By the DERPAR algorthm the solution diagram was calculated,which shows that the static bifurcation occurs under some conditions in the TPNC systems.Also,it shows that the static bifurcation occurs under some conditions in the TPNC systems.Also,it shows that,in a region of multiple solutions.the static instability may occur.It is defined as a region of thermal-siphon instability induced flow rate jumping.By means of the solution diagram,the stability margin can be determined in this region.Furthermore.the heat input at the peak of the solution diagram is defined as the maximum capacity of heating load that can be used to judge the capacity of the TPNC of a given geometry topological structure,Meanwhile,it is interesting that the TPNC systems have the hystersis phenomenon defined as thermal-siphon hysteresis.Some parametric effects related were also studied.     

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.     

An analytical investigation of role of expansion tank in semi-closed two-phase natural circulation loop

Role of the expansion tank in a semi-closed two-phase natural circulation loop was examined analytically with the emphasis placed on the flow instability. Loopwise steady circulation rate was obtained, and conditions for flow instability were examined by using the method of the linear stability analysis with perturbations. The homogeneous two-phase model was adopted for the analysis. As well as the pressure at the expansion tank, the length and the cross-sectional area of the tube connected to the expansion tank appeared to be the important parameters determining the flow instability. The system was predicted to be stable with the longer length and the smaller cross-sectional area of the expansion-tank line and also with the higher expansion-tank pressure.     

Non-dimensional analysis of static bifurcation in a natural circulation loop

The steady-state characteristics of a two-phase natural circulation loop were investigated based on the homogenous model. Transcendental equations of non-dimensional loop mass flow rate under various conditions were also derived. The static bifurcation diagram of a two-phase natural circulation described with non-dimensional variables Npch-m^＋ was obtained. In addition, various steady-state characteristics of a natural circulation loop were analyzed and discussed. These characteristics include the existence of multiple solutions under certain conditions, and the maximum mass flow rate. The authors also examined the effects of important parameters such as sub-cooling number, riser-to-heated-region length ratio, and riser-to-heated-region diameter ratio.     

An experimental study on natural circulation decay heat removal system for a loop type fast reactor

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.     

核动力装置强迫循环与自然循环过渡过程特性研究

针对某型压水堆核动力装置,建立反应堆及一回路系统强迫循环与自然循环的计算分析模型,并与试验值进行比较,验证了建立的模型计算精度高,满足工程分析的要求.利用建立的数学模型,对自然循环与强迫循环过渡过程进行分析计算,结果表明:强迫循环向自然循环转换过程中冷却剂流量、蒸汽发生器压力、反应堆出口温度是几个约束参数;自然循环向强迫循环的转换过程中反应堆功率变化与周期变化幅度较大.     

低流速自然循环过冷沸腾汽泡脱离点实验研究

针对低流速环形通道中自然循环过冷沸腾汽泡脱离点特性进行了实验研究,运用Saha-Zuber模型、Bowring模型、Forster模型和Levy模型测算汽泡脱离点的位置,与实验值进行比较,发现由于流速较低,上述模型不适用于计算本实验条件下的汽泡脱离点。以Levy模型为基础,修正了浮力项的影响系数,修正后的模型与实验结果符合良好。     

Analysis on natural circulation capacity of the CARR

The investigation on natural circulation （NC） characteristics of the China Advanced Research Reactor（CARR） is very valuable for practical engineering application and also a key subject for the CARR. In this study, a computer code was developed to calculate the NC capacity of the CARR under different pool water temperatures. Effects of the pool water temperature on NC characteristics were analyzed. The results show that with increasing pool water temperature, the NC flow rate increases while the NC capacity decreases. Based on the computation results and theoretical deduction, a correlation was proposed on predicting the relationship between the NC mass flow and the core power under different conditions. The correlation prediction agrees well with the computational result within ±10% for the maximal deviation. This work is instructive for the actual operation of the CARR.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

一维自然循环比例分析的结果与讨论

根据一维自然循环比例分析理论模型推导的试验装置与实际电站热工水力特性的相似准则,对整体性能试验装置主要参数的确定方法进行了深入讨论。结果表明:采用小尺度、等压力、同工质的实验装置模拟实际系统自然循环现象更为准确实际,单相和两相自然循环比例准则可同时满足,不存在复杂比例变化带来的失真,不利因素是试验成本偏高。同工质非等物性(不等压)模拟能够降低试验成本,但比例参数不能满足从单相自然循环到两相自然循环的平滑过渡。如保持功率连续,其速度比和特征时间比会有所差异。     

自然循环反应堆流量分配优化研究

由于自然循环反应堆一回路产生的驱动力有限,回路循环总流量较小,因此堆芯流量分配设计与优化非常重要。合理的堆芯流量分配不仅能满足热工安全要求,还能直接提高堆芯的性能。基于以上原因,本文对自然循环反应堆流量分配优化问题进行了初步研究,对闭式并联通道,采用一维流动传热模型,建立了入口阻力系数优化初值求解模型并设计了精确解搜索算法,并耦合堆芯热工分析程序COBRA编写了相应的堆芯流量分配优化程序。选择一自然循环反应堆算例,采用该程序对堆芯寿期内的流量分配优化进行了计算和分析。结果表明,将各典型寿期节点流量分配优化得到的入口阻力优化设置方案取平均值,可获得相对整个循环寿期达到较好优化效果的入口阻力设置方案。针对取平均值这种人工设计方法难以获得全局最优解的缺点,参考现代优化计算方法,提出了一种自动实现循环寿期内流量分配最优化的方法。     

非能动余热排出系统安全分析程序SAC-PREARS 的功能与应用

SAC-PREARS 是一个用于分析非能动RHRS稳态和瞬态安全特性的专用程序.通过实验验证的用于AC-600 非能动 RHRS安全分析的MISAP 程序,对SAC-PREARS程序进行了稳态计算验证.并应用SAC-PREARS程序对200 MW 核供热堆非能动RHRS稳态和瞬态热工水力特性进行了分析,得出了具有工程意义的结论.     

非能动余热排出系统数学模型研究与运行特性分析

利用某型核动力装置非能动余热排出系统1:10原理性试验的8个稳态工况、6个启动工况的试验数据,验证RELAP5/MOD3.2程序对本类型非能动余热排出系统的适用性。结果表明:垂直管内蒸汽凝结换热系数对两相流自然循环的流动与传热影响大;RELAP5/MOD3.2程序过低估算了垂直管内蒸汽流速对蒸汽凝结换热系数的影响,计算结果与试验结果偏差大。对RELAP5/MOD3.2程序垂直管内的蒸汽凝结换热模型进行修正,修正后的计算结果与试验值基本吻合;采用RELAP5程序对垂直管内两相流自然循环传热进行计算,须选择热前沿跟踪模型。对非能动余热排出系统的稳态与瞬态运行特性进行分析,理论计算与试验结果均表明:稳态工况下,系统可以实现稳定的两相流自然循环,系统排热能力受蒸汽发生器水位的影响大,冷却水入口温度与系统压力的影响相对较小;系统的启动特性良好,可快速地建立环路的自然循环,带走反应堆的衰变热。     

Sensitivity and uncertainty analysis for an accelerator-driven system with JENDL-4.0

A sensitivity and uncertainty analysis was performed for the accelerator-driven system (ADS) proposed by the Japan Atomic Energy Agency (JAEA) with the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0). Significant discrepancies have been found between the reactor physics parameters of JENDL-4.0 and those of JENDL-3.3. An analysis with the sensitivity coefficients showed that the major contributors to these discrepancies are the differences in the inelastic scattering cross sections of ²⁰⁶Pb and ²⁰⁷Pb, and the capture and inelastic scattering cross sections and ν value of ²⁴¹Am. The uncertainty analysis with the JENDL-4.0 covariance data found that the covariances of the fission neutron spectrum of minor actinides (MAs) have a considerable impact on the uncertainties of the reactor physics parameters.     

Development of three-dimensional reactor analysis code system for accelerator-driven system,ADS3D and its application with subcriticality adjustment mechanism

In order to perform the parametric survey for an accelerator-driven system (ADS) core with the subcriticality adjustment mechanism, a new calculation code system, ADS3D, was developed on MARBLE which is a comprehensive and versatile framework for reactor analysis. The application of ADS3D was also demonstrated on the neutronics design of ADS operated by control rod (CR) movement. Through the neutronics calculation, it was shown that the maximum proton beam current was decreased from 20.5 to 11.6 mA due to the switch from beam-operated to CR-operated core.